Analysis of 3-Dimensional Hydrogen Behavior With Passive Containment Cooling

2017 ◽  
Author(s):  
Xingguan Huang ◽  
Tingzao Fu ◽  
Gaofeng Huang
Keyword(s):  
Cooling System ◽  
Water Film ◽  
Severe Accident ◽  
Accident Analysis ◽  
3 Dimensional ◽  
Film Evaporation ◽  
Lumped Parameter ◽  
Development Procedure ◽  
Hydrogen Analysis ◽  
Hydrogen Safety

Hydrogen safety is one of the most important issues for severe accident analysis. Comparing with the lumped parameter code, 3-D CFD code GASFLOW has more advantages for analyzing hydrogen related phenomena [1]. However, due to the lacking of passive containment cooling system (PCS) model, GASFLOW is inapplicable for hydrogen analysis of PWR with PCS. This paper shows the development procedure of PCS model, including the model introduction and validation. The main functions of the PCS model are the calculation of falling water film thickness, heat transfer between wall and film, and film evaporation. Then, the hydrogen safety analysis by GASFLOW with PCS model is performed for CAP1400.

Preliminary Severe Accident Analysis of Fusion-Fission Hybrid Reactor Using the MELCOR Code

2014 ◽  
Author(s):  
Zhang Dabin ◽  
Zhiwei Zhou ◽  
Heng Xie ◽  
Tang Yang
Keyword(s):  
Cooling System ◽  
Safety Performance ◽  
Severe Accident ◽  
Hybrid Reactor ◽  
Accident Analysis ◽  
Analysis Model ◽  
Decay Heat ◽  
Calculation Results ◽  
Core Meltdown ◽  
Dry Out

The fusion-fission hybrid conceptual reactor is a proposed means of generating power, which adopts a water cooled fission blanket based on ITER. Due to the water cooled fission blanket, safety performance of the hybrid reactor should be considered, including decay heat remove, core uncovered, core meltdown, core degradation, radioactivity releases, etc., similar with the PWRs. The main goal of this work is to develop the fission blanket model by using MELCOR code, and to evaluate the safety performance under severe accidents preliminarily. Based on MELCOR 1.8.5, some modification is made for the severe accident analysis of fission blanket. Using modified MELCOR code, an analysis model is developed for the fission blanket and the cooling loop. The strategy of the In-Vessel Retention (IVR) using the ex-vessel cooling method is evaluated during a large break LOCA. The calculation results describes the main phenomena during the severe accident progression, including core dry out, meltdown, relocation, etc.. Simulation result is shown that the decay heat in the fission zone can be removed out by the ex-vessel cooling system merely, and the fuel max temperature will not reach the melting point.

Accident Analysis of Fukushima Daiichi Nuclear Power Plant Unit 2 by the SAMPSON Severe Accident Code

2014 ◽  
Author(s):  
Atsuo Takahashi ◽  
Marco Pellegrini ◽  
Hideo Mizouchi ◽  
Hiroaki Suzuki ◽  
Masanori Naitoh
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Accident Analysis ◽  
Unknown Boundary ◽  
Long Time ◽  
Fukushima Daiichi ◽  
Nuclear Power Plant Unit

The transient process of the accident at the Fukushima Daiichi Nuclear Power Plant Unit 2 was analyzed by the severe accident analysis code, SAMPSON. One of the characteristic phenomena in Unit 2 is that the reactor core isolation cooling system (RCIC) worked for an unexpectedly long time (about 70 h) without batteries and consequently core damage was delayed when compared to Units 1 and 3. The mechanism of how the RCIC worked such a long time is thought to be due to balance between injected water from the RCIC pump and the supplied mixture of steam and water sent to the RCIC turbine. To confirm the RCIC working conditions and reproduce the measured plant properties, such as pressure and water level in the pressure vessel, we introduced a two-phase turbine driven pump model into SAMPSON. In the model, mass flow rate of water injected by the RCIC was calculated through turbine efficiency degradation the originated from the mixture of steam and water flowing to the RCIC turbine. To reproduce the drywell pressure, we assumed that the torus room was flooded by the tsunami and heat was removed from the suppression chamber to the sea water. Although uncertainties, mainly regarding behavior of debris, still remain because of unknown boundary conditions, such as alternative water injection by fire trucks, simulation results by SAMPSON agreed well with the measured values for several days after the scram.

scholarly journals Numerical Investigation of Water Film Evaporation with the Countercurrent Air in the Asymmetric Heating Rectangular Channel for Passive Containment Cooling System

2020 ◽  
Vol 2020 ◽  
pp. 1-17
Author(s):  
Kashuai Du ◽  
Po Hu ◽  
Zhen Hu
Keyword(s):  
Heat Transfer ◽  
Numerical Simulation ◽  
Cooling System ◽  
Rectangular Channel ◽  
Water Film ◽  
Annular Channel ◽  
Countercurrent Flow ◽  
Film Evaporation ◽  
Air Inlet ◽  
Asymmetric Heating

Passive containment cooling system (PCCS) is an important passive safety facility in the large advanced pressurized water reactor. Using the physical laws, such as gravity and buoyancy, the water film/air countercurrent flow is formed in the external annular channel to keep inside temperature and pressure below the maximum design values. Due to the large curvature radius of the annular channel, one of the short arc segments is taken out, as a rectangular channel, to analyze the main water film evaporation heat transfer characteristics. Two numerical methods are used to predict the water film evaporative mass flow rate during the heat transfer process in the large-scale rectangular channel with asymmetric heating when the water film temperature is not saturated. At the same time, these numerical simulation results are validated by the experiment which is set up to study water film/air countercurrent flow heat transfer on a vertical back heating plate with 5 m in length and 1.2 m in width. It is shown that the maximum deviation between numerical simulation and experiment is 30%. In addition, the influences on these parameters, such as heat flux, evaporative mass flow rate, and water film thickness, are evaluated under the different tilted angles of the rectangular channel and horizontal plane, water/air inlet flow rates, water/air inlet temperatures, heating surface temperatures, and air inlet relative humidities. All these results can provide a good guidance for the design of PCCS in the future.

scholarly journals MELCOR Modeling of Combined Accident Tolerant Fuel and Reactor Core Isolation Cooling System Operation

2020 ◽  
Author(s):  
Chris Faucett ◽  
Bradley Beeny ◽  
Karen Vierow Kirkland
Keyword(s):  
Hydrogen Production ◽  
Cooling System ◽  
Reactor Core ◽  
The United States ◽  
New Physics ◽  
Department Of Energy ◽  
Severe Accident ◽  
United States Government ◽  
Accident Analysis ◽  
Combined Use

Abstract The work presented in this paper presents new techniques for modeling the combined use of the Reactor Core Isolation Cooling (RCIC) System and Accident Tolerant Fuel (ATF) in a Boiling Water Reactor (BWR). With guidance from Sandia National Laboratories’ Severe Accident Analysis department, a MELCOR BWR model was developed from open source literature. The demonstration shown herein simulates BWR long-term station blackout (LTSBO) conditions with the Nuclear Regulatory Commission’s (NRC) MELCOR severe accident analysis code. By combining state-of-the-art MELCOR modeling practices with new, physics-based RCIC System and ATF MELCOR inputs, this BWR model provides a contemporary platform for BWR severe accident simulations. The authors are investigating the combined use of the RCIC System and ATF as a means of passively enhancing reactor safety. The benefits of this approach were evaluated by performing simulations using traditional fuel designs (i.e. Zircaloy cladding) and ATF with an iron-chromium-aluminum (FeCrAl) clad under BWR LTSBO conditions. ATF performance was evaluated using severe accident metrics, specifically event sequence timings and the hydrogen production rate from cladding oxidation. Preliminary results show delayed core degradation timelines with less hydrogen production for ATF simulations. Although the results are limited in scope, the presented analysis could easily be expanded to a full-scale uncertainty study that considers a range of severe accident boundary conditions. This paper describes objective technical results and analysis. Any subjective views or opinions that might be expressed in the paper do not necessarily represent the views of the U.S. Department of Energy or the United States Government.

A New Model Towards Thermal Mixing and Stratification in Passive Containment

2013 ◽  
Author(s):  
He Zhang ◽  
Fenglei Niu ◽  
Yu Yu ◽  
Peipei Chen
Keyword(s):  
Heat Transfer ◽  
System Analysis ◽  
Cooling System ◽  
Transfer Problem ◽  
Ambient Fluid ◽  
Simulation Code ◽  
Thermal Mixing ◽  
Lumped Parameter ◽  
Passive Safety System ◽  
Transfer Problems

Thermal mixing and stratification often appears in passive containment cooling system (PCCS), which is an important part of passive safety system. So, it is important to accurately predict the temperature and density distributions both for design optimization and accident analysis. However, current major reactor system analysis codes only provide lumped parameter models which can only get very approximate results. The traditional 2-D or 3-D CFD methods require very long simulation time, and it’s not easy to get result. This paper adopts a new simulation code, which can be used to calculate heat transfer problems in large enclosures. The new code simulates the ambient fluid and jets with different models. For the ambient fluid, it uses a one-dimensional model, which is based on the thermal stratification and derived from three conservation equations. While for different jets, the new code contains several jet models to fully simulate the different break types in containment. Now, the new code can only simulate rectangular enclosures, not the cylinder enclosure. So it is meaningful for us to modify the code to simulate the actual containment, then it can be applied to solve the heat transfer problem in PCCS accurately.

Combined Experimental and CFD Study of a HP Blade Multi-Pass Cooling System

2009 ◽  
Author(s):  
D. Jackson ◽  
P. Ireland ◽  
B. Cheong
Keyword(s):  
Heat Transfer ◽  
Film Cooling ◽  
Cooling System ◽  
Detailed Comparison ◽  
Bulk Temperature ◽  
Internal Cooling ◽  
3 Dimensional ◽  
Turbine Cooling ◽  
Cooling Hole ◽  
The Difference

Progress in the computing power available for CFD predictions now means that full geometry, 3 dimensional predictions are now routinely used in internal cooling system design. This paper reports recent work at Rolls-Royce which has compared the flow and htc predictions in a modern HP turbine cooling system to experiments. The triple pass cooling system includes film cooling vents and inclined ribs. The high resolution heat transfer experiments show that different cooling performance features are predicted with different levels of fidelity by the CFD. The research also revealed the sensitivity of the prediction to accurate modelling of the film cooling hole discharge coefficients and a detailed comparison of the authors’ computer predictions to data available in the literature is reported. Mixed bulk temperature is frequently used in the determination of heat transfer coefficient from experimental data. The current CFD data is used to compare the mixed bulk temperature to the duct centreline temperature. The latter is measured experimentally and the effect of the difference between mixed bulk and centreline temperature is considered in detail.

scholarly journals Evaluation of passive autocatalytic recombiners operation efficiency by means of the lumped parameter approach

Nukleonika ◽  
2015 ◽  
Vol 60 (2) ◽  
pp. 339-345 ◽  
Author(s):  
Tomasz Bury
Keyword(s):  
Hydrogen Generation ◽  
Cooling System ◽  
Component Mixture ◽  
Pressurized Water ◽  
Water Reactor ◽  
Safety Issues ◽  
Loss Of Coolant Accident ◽  
Lumped Parameter ◽  
Core Cooling ◽  
Parameter Approach

Abstract The problem of hydrogen behavior in containment buildings of nuclear reactors belongs to thermal-hydraulic area. Taking into account the size of systems under consideration and, first of all, safety issues, such type of analyses cannot be done by means of full-scale experiments. Therefore, mathematical modeling and numerical simulations are widely used for these purposes. A lumped parameter approach based code HEPCAL has been elaborated in the Institute of Thermal Technology of the Silesian University of Technology for simulations of pressurized water reactor containment transient response. The VVER-440/213 and European pressurised water reactor (EPR) reactors containments are the subjects of analysis within the framework of this paper. Simulations have been realized for the loss-of-coolant accident scenarios with emergency core cooling system failure. These scenarios include core overheating and hydrogen generation. Passive autocatalytic recombiners installed for removal of hydrogen has been taken into account. The operational efficiency of the hydrogen removal system has been evaluated by comparing with an actual hydrogen concentration and flammability limit. This limit has been determined for the three-component mixture of air, steam and hydrogen. Some problems related to the lumped parameter approach application have been also identified.

scholarly journals Severe Accident Analysis Code SAMPSON Improvement for IMPACT Project.

2002 ◽  
Vol 45 (3) ◽  
pp. 607-614 ◽  
Author(s):  
Hiroshi UJITA ◽  
Takashi IKEDA ◽  
Masanori NAITOH
Keyword(s):  
Severe Accident ◽  
Accident Analysis

Validation of Film Evaporation Model in GASFLOW-MPI

2018 ◽  
Author(s):  
Li Yabing ◽  
Zhang Han ◽  
Xiao Jianjun
Keyword(s):  
Experimental Data ◽  
Wall Temperature ◽  
Cooling System ◽  
Numerical Study ◽  
Thermal Power ◽  
Heat Removal ◽  
Air Velocity ◽  
Evaporation Model ◽  
Film Evaporation ◽  
Test Region

A dynamic film model is developed in the parallel CFD code GASFLOW-MPI for passive containment cooling system (PCCS) utilized in nuclear power plant like AP1000 and CAP1400. GASFLOW-MPI is a widely validated parallel CDF code and has been applied to containment thermal hydraulics safety analysis for different types of reactors. The essential issue for PCCS is the heat removal capability. Research shows that film evaporation contributes most to the heat removal capability for PCCS. In this study, the film evaporation model is validated with separate effect test conducted on the EFFE facility by Pisa University. The test region is a rectangle gap with 0.1m width, 2m length, and 0.6m depth. The water film flowing from the top of the gap is heated by a heating plate with constant temperature and cooled by countercurrent air flow at the same time. The test region model is built and analyzed, through which the total thermal power and evaporation rate are obtained to compare with experimental data. Numerical result shows good agreement with the experimental data. Besides, the influence of air velocity, wall temperature and gap widths are discussed in our study. Result shows that, the film evaporation has a positive correlation with air velocity, wall temperature and gap width. This study can be fundamental for our further numerical study on PCCS.

Sign in / Sign up

Export Citation Format

Share Document